Many nuclear reactors are typically constructed as boiling water reactors where suitable nuclear fuel is disposed in a reactor pressure vessel in which water is heated. The water and steam are carried through various components and piping which are typically formed of stainless steel, with other materials such as alloy 182 weld metal and alloy 600 used for various components directly inside the reactor pressure vessel.
Materials in the reactor core region are susceptible to irradiation assisted stress corrosion cracking. This is because the material in the core region is exposed to the highly oxidizing species generated by the radiolysis of water by both gamma and neutron radiation under normal water chemistry conditions, in addition to the effect of direct irradiation assisted stress corrosion cracking. The oxidizing species increases the electrochemical corrosion potential of the material, which in turn increases its propensity to undergo intergranular stress corrosion cracking or irradiation assisted stress corrosion cracking.
Suppression of the oxidizing species carried within such materials is desirable in controlling intergranular stress corrosion cracking. An effective method of suppressing the oxidizing species coming into contact with the material is to inject hydrogen into the reactor water via the feedwater system so that recombination of the oxidants with hydrogen occurs within the reactor core.
This method is called hydrogen water chemistry and is widely practiced for mitigating intergranular stress corrosion cracking of materials in boiling water reactors. When hydrogen water chemistry is practiced in a boiling water reactor, the electrochemical corrosion potential of the stainless steel material decreases from a positive value generally in the range of 0.060 to 0.200 V (Standard Hydrogen Electrode Potential, or “SHE”) under normal water chemistry to a value less than −0.230 (SHE). When the electrochemical corrosion potential is below this negative value, intergranular stress corrosion cracking of stainless steel can be mitigated and its initiation can be prevented.
Considerable efforts have been made in the past decade to develop reliable electrochemical corrosion potential sensors to be used as reference electrodes that can determine the electrochemical corrosion potential of operating surfaces of reactor components.
The typical electrochemical corrosion potential sensor experiences a severe environment in a view of the temperature of the water wall exceeding 288° C.; relatively high flow rates of the water up to and exceeding several m/s; and the high nuclear radiation in the core region.
A drawback of currently available sensors is that they have a limited lifetime in that some have failed after only three months of use while a few have shown evidence of operation for approximately six to nine months. Since many of the locations where these sensors are installed are inaccessible during plant operation, a lifetime of at least 24 months must be obtained to allow for continuous monitoring during the entire fuel cycle (fuel cycle length is plant-specific but can range between 12 months to 24 months).
The invention disclosed here is most closely associated with commonly owned U.S. Pat. Nos. 5,043,053 and 6,357,284. Of the various types of sensors currently available, one (as disclosed in the '053 patent) includes a ceramic probe packed with a mixture of metal and metal oxide powder. Predominant failure modes for this type of ECP sensor include degradation of the ceramic material (yttria partially-stabilized zirconia) and cracking and corrosive attack of, the ceramic-to-metal braze used to attach the ceramic material to the balance of the sensor assembly.
Accordingly, it is desired to develop an improved ceramic electrochemical corrosion potential sensor addressing the insufficient useful life.